ML082520013

University of Missouri, Request to Amend Technical Specifications

ML082520013
Person / Time
Site:	University of Missouri-Columbia
Issue date:	09/04/2008
From:	Rhonda Butler Univ of Missouri - Columbia
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML082520013 (11)

Text

Research Reactor Center University of Missouri-Columbia 1513 Research Park Drive Columbia, MO 65211 PHONE 573-882-4211 FAX 573-882-6360 wEB http://web.missouri.edu/-murrwww September 4, 2008 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Mail Station P 1-37 Washington, DC 20555-0001

Reference:

Docket 50-186 University of Missouri - Columbia Research Reactor Amended Facility License R-103 On June 8, 2007, the University of Missouri - Columbia Research Reactor submitted a request to amend the Technical Specifications appended to Facility License R-103.

Enclosed is our response to the U.S. Nuclear Regulatory Commission's request for additional information regarding the proposed amendment, dated August 26, 2008.

If you have any questions, please contact Leslie P. Foyto, the facility Reactor Manager, at (573) 882-5276 or foytol Rcmissouri.edu.

Sincerely, Ralph A. Butler, P.E.

Director zlý2- *I RAB/djr Enclosures

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MARGEE P. STOWT My CmnuDSm~ EOMre Cwsew #08511436

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Research Reactor Center University of Missouri-Columbia September 4, 2008 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Mail Station P 1-37 Washington, DC 20555-0001 1513 Research Park Drive Columbia, MO 65211 PHONE 573-882-4211 FAX 573-882-6360 WEB http://web.missouri.edu/-murrwww

SUBJECT:

Docket 50-186 University of Missouri - Columbia Research Reactor Amended Facility License R-103 Written communication as specified by 10 CFR 50.4(b)(1) regarding the response to the "University of Missouri at Columbia - Request for Additional Information Re: License Amendment on Fueled Experiment Conditions (TAC No. MD5782)," dated August 26, 2008 By letter dated June 8, 2007, the University of Missouri - Columbia Research Reactor (MURR) submitted a request to the U.S. Nuclear Regulatory Commission (NRC) to amend the Technical Specifications, which are appended to Facility License R-103, in order to perform an experiment in support of a U.S. Department of Energy (DOE) program to demonstrate the feasibility of producing fission product molybdenum-99 (Mo-99) using low-enriched uranium (LEU) foil targets.

On August 10, 2007, the NRC requested additional information and clarification regarding the proposed amendment in the form of four (4) questions. By letter dated January 10, 2008, the MURR responded to those questions.

On March 19, 2008, the NRC requested additional information and clarification regarding the proposed amendment and the responses to the initial request for additional information in the form of one (1) question. By letter dated April 15, 2008, the MURR responded to that question.. On August 26, 2008, the NRC requested additional information and clarification regarding the proposed license amendment in the form of ten (10) questions. Those questions, and MURR's responses to those questions, are attached.

If there are questions regarding this response, please contact me at (573) 882-5276. I declare under penalty of perjury that the foregoing is true and correct.

ENDORSEMENT:

Sincerely, Reviewed and Approved, Leslie P. Foyto Reactor Manager 21 A, `ý45 ---

I -ý-Zl --4-5 Ralph A. Butler, P.E.

Director MARGEE P. STOUT N W March 24,2012

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Reactor Advisory Committee Reactor Safety Subcommittee Dr. Robert Duncan, Vice Chancellor for Research Mr. Craig Basset, U.S. NRC Mr. Alexander Adams, U.S. NRC

1.

Your response to RAI 4 dated January 10, 2008, described the six major steps in the LEU-Modified Cintichem process. Your responses to RAI I dated January 10, 2008, indicated that your quality control checks, including leak testing of the seal welded container, will be implemented prior to irradiation. Is the first step, annular target fabrication, going to be perfbrmed at MURR in Columbia, Missouri or by others at an offsite facility? If it is at an offsite.flbrication facility, what quality assurance program is in place to control these activities ? If offsite, briefly describe your planned periodic audit of the fabricator's quality assurance program. If no audit is planned, please justify. Describe the receipt inspection procedure for receiving nuclear materials, such as the encapsulated targets, at your facility.

All types of experiments conducted at MURR must be reviewed and approved by the Reactor and Reactor Health Physics Managers. The mechanism for obtaining such approval is a Reactor Utilization Request (RUR).

The RUR describes the experiment in considerable detail.

It presents the activities and isotopes which are produced and details the methods for post-irradiation handling. The most important section of the RUR, and one which is given paramount consideration in its preparation, is the safety analysis. The safety analysis includes all credible accident and transient scenarios to ensure that the experiment does not jeopardize the safe operation of the reactor or constitute a hazard to the safety of the facility staff and general public.

A 50.59 Screen is also prepared with each RUR to demonstrate and document that the safety analysis meets the requirements of all applicable licensing basis documents, including the MURR Hazards Summary Report and Technical Specifications.

The RUR that has been prepared, reviewed and approved for this experiment is RUR 431, "Uranium, Low Enriched Uranium Foil" [1]. The RUR was reviewed and approved on June 25, 2008 by the Reactor Manager and then subsequently reviewed by the Reactor Safety Subcommittee on July 2, 2008.

The LEU-foil target assembly/fabrication procedure referenced in RUR 431 was provided by Argonne National Laboratory (ANL). ANL has significant experience in fabricating/assembling annular LEU-foil targets. For the first target, partial assembly, not including welding and leak checking, was performed at ANL, Argonne, Illinois, in the presence of the MURR Reactor Manager.

It was decided that this approach would be beneficial and appropriate because of ANL's assembly experience and because of the very limited number of targets we intend to fabricate for this proof of principle experiment. No other offsite fabrication facilities will be involved in the fabrication/assembly of our targets.

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Partial target assembly included verification of all materials, including their weights, and cleanliness of the target assembly equipment. ANL also performed the draw die process, which effectively swages the inner and outer aluminum tubes together, thus enveloping the LEU and nickel foils within the tubes. The target was then placed in a helium atmosphere, sealed and shipped to MURR.

The target was welded and leak tested at MURR in accordance with procedure IRR-PSO-105 (Rev. 3), "Encapsulation Leak Check."

MURR has in place strict inspection and receipt guidelines for receiving fabricated HEU fuel assemblies and these readily extend to the receipt of LEU-foil targets. The existing fuel receipt procedure, as outlined in the "MURR Special Nuclear Materials Manual," implements all applicable NRC regulations regarding the transfer and accountability of Special Nuclear Material (SNM).

Any subsequent target fabrication will be made at MURR using the procedure provided by ANL in the presence of the MURR Reactor Manager. Any larger targets, or targets that will generate a fission product inventory greater than what is specified in our current license amendment request, will require an additional amendment request to the NRC.

2.

The radiological calculations presented in your responses to RAI I and 3 dated January 10, 2008, are based on a 5-gram LEU target and the radioiodine and noble gas activities

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produced by irradiation at a thermal neutron flux of 1.5 x 10 n/cm -sec for 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br />.

Describe the tolerances specified in the experimental plan for the LEU target mass, the thermal neutron fluence, and irradiation time to validate the assumptions in the radiological calculation?

By using an LEU-foil mass of less than 5 grams, irradiating the target in a location where the thermal flux has been mapped and shown to be on average 1.5 x 10+13 n/cm 2-sec or lower, and irradiating the target for less than 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br />, the radioiodine and noble gases activities in the target will be no greater than the values listed in our response, dated January 10, 2008, to the August 10, 2007 RAI. All three of these parameters collectively determine the fission product inventory in the target and set the safety envelope for which the experiment will be run in.

The weight of each LEU-foil will be measured prior to target assembly using a calibrated scale, the flux in the irradiation position will be measured to the precision that our methods allow

(+/-10%), and the irradiation time will be controlled as required.

The combination of fissile material mass, thermal flux and run time will be limited such that the limits of MURR Technical Specification 3.6.a are not exceeded.

3.

The radiological calculations presented in your responses to RAI I and 3 dated January 10, 2008, are based on assuming a 2 minute total evacuation time for the containment building following actuation of the isolation system.

Are periodic evacuation drills 3 of 10

conducted to verify this? Are all unescorted persons inside the containment required to have periodic evacuation training as one of the requirements for unescorted access?

As required by the MURR Emergency Plan and Implementing Procedures, an on-site facility-wide emergency drill is conducted annually. As part of this annual emergency drill, a facility evacuation, which includes a reactor isolation, is performed. The reactor isolation tests the ability of the staff to evacuate the containment building within 2 minutes from actuation of the isolation system. All personnel, whether or not they are escorted or unescorted, are required to evacuate during this time interval.

Emergency preparedness training requirements are outlined in procedure EP-RO-003 (Rev. 2),

"Emergency Preparedness Training."

Initial and annual training on radiation safety and the emergency plan and implementing procedures, which includes a person's actions during both a reactor isolation and facility evacuation, is conducted for anyone that has unescorted access to the facility. The training emphasizes each person's role during an emergency, specifically a reactor isolation and facility evacuation.

4.

The radiological calculations presented in your responses to RAL 1 and 3 dated January 10, 2008, determine the radiological impact to members of the public from potential failure of the fueled experiment radiation container. These calculations are sensitive to the building ventilation system exhaust stack release flow rate. Is the exhaust stack flow monitored periodically to validate the assumptions in the radiological calculation?

Flow rate (in standard cubic feet per minute) through the facility ventilation exhaust stack is measured biannually, via an exhaust plenum test port adjacent to the stack monitor isokentic probe, using a velocity probe and ADM-870 Airdata Multimeter.

The stack monitor continuously measures the airborne concentrations of radioactive particulate, iodine, and noble gases in the facility exhaust air. As required by the MURR Technical Specifications, the monitor is calibrated semi-annually to assure a consistent response to the radionuclides expected during routine reactor operation or a hypothetical accidental release. Additionally, a magnehlic gauge, which continually measures differential pressure in the exhaust plenum, is installed as a secondary means of verifying this flow rate.

5.

Your amendment request letter dated June 7, 2007, stated that the encapsulated LEU-foil target will be held in place in the irradiation position by a sample-handling device and flux mapping of the position will be performed prior to conducting the experiment. Describe the procedure(s) to accomplish flux mapping, fueled experiment target handling, and securing the target in place in the irradiation position. Have these tools and handling devices been used for other experiments at the MURR or will they be developed especially for this experiment? What are the training and qualifications for personnel authorized to perJorm this work?

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Flux mapping of the graphite reflector region is periodically performed in order to characterize and maintain an up-to-date flux profile of the various irradiation positions. For this specific experiment, the flux profile was measured on three separate occasions using a flux wire can fabricated to essentially the same physical dimensions as the actual LEU-foil target. Each can contained three (3) dilute cobalt-in-aluminum alloy flux wires (characterized for its cobalt content traceable to NIST through NIST SRM 953 "Neutron Density Monitor Wire"), spaced 1200 apart, that extended above and below the 1-inch vertical LEU-foil dimension. The flux wire can was then placed in the K2 position, the pre-selected irradiation position for this experiment, at 10 MW such that the flux wires were positioned at the same height that the LEU-foil target will be positioned at during irradiation. After sufficient activation had been obtained (typically 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />), the flux wires were removed, allowed to decay, segmented and counted, and analyzed to determine the mean thermal flux.

With the exception of the sample-handling device that will position the LEU-foil target at the correct height in the K2 irradiation position, all of the other sample handling equipment needed for this experiment are time tested and routinely used on other experiments.

The sample-handling device has been specifically designed for this target and has already been used on four separate occasions: once to test the thermocouple attachment mechanism on a "dummy" target (which was in place for approximately 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> during 10 MW operation) and three times to measure the flux profile.

The annular LEU-foil target will be placed in the sample-handling device, then the device and target lowered into the irradiation position using a handling cable. The experiment will initially be classified and treated as an "Unsecured Experiment," as defined by MURR Technical Specification 1.25, until the reactivity worth of the experiment is measured. Once the reactivity measurement is performed and verified to be less than that of a "Movable Experiment," as defined by MURR Technical Specification 1.11, the experiment can be moved into and out of the irradiation position while the reactor is operating. Additionally, an MCNP calculation will be performed in order to estimate the reactivity worth of the target prior to irradiation.

Once the irradiation is complete, the sample-handling device will be removed from the irradiation position and allowed to decay a minimum of 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />. The LEU-foil target will then be removed from the sample-handling device using an underwater manipulator and placed in a sample cask, which has been specifically selected because of its shielding qualities, and transported to the hot cell.

Due to the minimal number of targets that will be irradiated under the experiment plan, the Reactor Manager, with the assistance of licensed reactor operators as needed, will perform all of the in-pool work. Practice and experience that were gained while handling the "dummy" target and the three flux wire cans can be transferred to the handling of the actual LEU-foil target.

6.

Your amendment request letter dated June 7, 2007, stated the reactivity worth of encapsulated LEU-foil target will be measured prior to an irradiation run to assure the 5 of 10

limits stated in the Technical Specification (TS) 3.1.j are not exceeded. Describe the procedure for measuring the reactivity worth of the LEU-foil target. Will this be measured be/bre each irradiation run? Will only one LEU-foil target be irradiated during a reactor operating period? Do you expect any significant change in reactivity worth of the encapsulated LEU-fbil target during an experiment run?

The procedure that will be used to measure the reactivity worth of the LEU-foil target is RP-RO-201 (Rev. 2), "Measurement of Reactivity Worth of Flux Trap Loadings or Individual Samples, RTP-17(B)."

The methodology used by this procedure employs the known, measured integral control rod/bank worth to estimate the reactivity effect of an unknown sample by performing two successive low power startups - one with the sample (target) installed and one without. The difference in the control rod/bank height is then used to determine the reactivity worth of the target after correcting for any temperature effects.

Once this method is used to verify that the target can be classified as a "Movable Experiment," the target will then either be inserted or withdrawn, depending on it's reactivity worth (positive or negative). A more precise reactivity worth can then be calculated from the measured doubling time.

Additionally, an MCNP calculation will be performed in order to estimate the reactivity worth of the target prior to irradiation.

We expect the reactivity effect of the 5-gram LEU-foil target to be minimal since the irradiation will be performed in the graphite reflector region of the reactor where the sample is, for the most part, decoupled from the core by the beryllium reflector that is situated between the outer pressure vessel and the graphite reflector irradiation positions.

Should we irradiate more than one LEU-foil target as part of this feasibility study, we will perform a reactivity measurement of each target.

From a reactor safety perspective, any significant deviation from the expected reactivity of a target will be captured by the tight tolerances on the Estimated Critical Position (ECP) limits that are imposed during a reactor startup.

We do not expect any significant change in the reactivity worth of the LEU-foil target during the experiment run.

7.

TS 3.6.n limits the maximum temperature of fjeled experiments to at least a factor of two below the melting temperature of any material in the experiment and first-of-a-kind experiments will be instrumented to measure temperature. Has this temperature been determined.for the encapsulated LEU-foil target experiment? If so, briefly describe the methodology used.

Describe the instrumentation to measure the temperature of the irradiation container.

As part of the safety analysis thermal evaluation documented in RUR 43 1, the maximum internal temperature of the LEU-foil target was calculated using the computer code HEATING 7.2.C and 6 of 10

a 2-dimensional R-Z model of the target geometry.

After performing multiple thickness measurements of the LEU-foil (maximum thickness measured was 0.008-inches), an LEU-foil thickness of 0.010 inches (0.026 cm) was conservatively used in the R-Z model. An air gap of 0.001 cm was also conservatively modeled on each side of the 0.002 cm thick nickel foil. The nickel foil, which surrounds the LEU foil, is used as a recoil barrier whereas the air gaps were included to account for any thermal contact resistance. The target parameters for density and thermal conductivity used to perform the calculations were standard values listed in the tables built into the computer code. The maximum internal temperature calculated by the HEATING code is 200.70 C. The maximum calculated surface temperature is 100.5' C.

An LEU-foil target consists of the following three materials: aluminum, nickel and uranium.

The melting point of these materials are 6600 C for aluminum, 14550 C for nickel and 11350 C for uranium.

Therefore, as required by MURR Technical Specification 3.6.n, the maximum temperature limit on the LEU-foil target during irradiation is 330' C (1/2 the melting temperature of aluminum).

The calculated heat flux of the LEU-foil is 19.3 w/cm2, which is well below the administrative limit of 38 w/cm 2 for reflector region experiments. The calculated internal temperature of 200.7' C is well below the maximum temperature limit on. the target. Additionally, the calculated surface temperature of 100.50 C indicates that the target will not exceed the saturation temperature of the cooling medium.

The target will be instrumented with no less than three (3) K-type thermocouples connected to a Honeywell Minitrend QX Graphic Recorder. The recorder will provide a continuous display of all thermocouple outputs in degrees Fahrenheit.

The recorder will also provide visual and audible "Alert" and "Alarm" functions on high temperature. The thermocouples will be attached to the LEU-foil target using epoxy compound Duralco 4700 and 25 gauge surgical stainless steel wire. As mentioned in the response to question 5, this attachment mechanism was satisfactorily tested at 10 MW operation at the same flux and fluence that the actual LEU-foil target will be irradiated to.

8.

Your response to RAI1 dated January 10, 2008, stated that the proposed experiment was calculated to produce a heat flux of approximately 19.5 W/cm 2 under the assumed operating conditions, which is about half of the administrative limit. What parameters determined the establishment of the facility administrative heat flux limit of 38 W/cm2 ?

The MURR Hazards Summary Report allows the facility to use precedence [2] from previous experiments in our safety analysis of new experiments.

The administrative limit placed on experiments in the graphite reflector region was derived from very early empirical measurements, including a calorimeter measurement. This limit has been in effect for over 30 years and has been found to be very safe and effective. We have never had an experiment, even those whose heat flux approached the administrative limit, fail over that entire period of time or fail to meet any MURR Technical Specification related to experiment temperature.

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For this experiment, the heat flux of the LEU-foil has been calculated to be about half of this time-tested administrative limit.

9.

The proposed TS 3.6.0 amendment refers to the pressure specification in TS 3.6.i for irradiation containers. Briefly describe how the maximum internal pressure for the irradiation container for the LEU-foil target experiment was determined for the highest temperature predicted during the irradiation run?

As part of the safety analysis decomposition/pressurization evaluation documented in RUR 431, the maximum internal pressure of the LEU-foil target was calculated as follows.

The LEU and nickel foils are encapsulated in a sealed aluminum annulus vessel. This target and encapsulation design has been utilized at reactors in Australia, Argentina, and Indonesia where no reported incidences of failure due to pressurization have occurred. No swelling of the target or encapsulation is expected and no pressurization or mechanical stresses on the encapsulations are anticipated. Any pressurization would be determined by the void volume within the target, which is minimized by the fabrication process. The tubing which holds the LEU and nickel foil envelope is mechanically swaged in order to maximize the physical contact of the aluminum tubes with each other and with the foil envelope so that any internal void volume is essentially eliminated. Therefore, no measureable void volume should exist in the encapsulation at the end of fabrication. Any pressurization would be from the temperature rise in the minute contained void volume during fabrication and the slight amount of fission product gasses produced at the surface of the uranium foil during irradiation.

For the purposes of this evaluation, it is conservatively assumed that iodine will behave as and be considered a fission product gas. Irradiation of the LEU-foil target will produce fissioning and resultant fission product gasses, including both stable and radioactive isotopes of krypton, xenon, and iodine. The mass of potential gasses which could contribute to pressurization is finite and in this case, miniscule.

The inventory of fission product gasses has been calculated using the computer code ORIGEN.

This total inventory of potential gas is 6.79E-6 moles.

It is also assumed that any pre-existing void space would be at atmospheric pressure (14.7 psi at 270 C).

Applying some of the methodology [3,4] used by the Australian Nuclear Science and Technology Organization (ANSTO), a conservative hypothetical estimate of pressurization using the Ideal Gas Equation, P = nRT/V, was calculated. The amount of gas available is a finite volume and the smallest void volume will produce the greatest pressure potential.

A void volume equivalent to a 0.01 mm gap covering an area the size of the LEU-foil was assumed, therefore V= 0.0195 cm 3 or 1.95E-5 liters or 7.92E-7 moles of gas.

Making the further assumptions that (1) T = 3300 C, the maximum operating temperature limit of the LEU-foill('

(1.65 times the maximum internal temperature calculated by MURR using the HEATING code),

(2) 6% of the fission product gas is released into the void, and (3) the potential gasses (including iodine) behave as ideal gasses, pressure was calculated as follows:

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P (psia) =

((6.79E-6 moles x 0.06) + (7.92E-7 moles)) (0.0821 ltr*atmr!K-mole) (603* K)

(14.7 psi/atm) / 1.95E-5 liters P (psia) =

45 MURR Technical Data Report No. 0112 [5] assessed the potential fission gas release during LEU-foil target irradiation and predicts that less than 6% of the available gas atoms have the potential to escape the boundaries of the uranium foil, and that even the gas that might escape would be captured and held by the nickel foil recoil barrier.

However, even if we assume a worst case scenario of a 100% release of fission product gases into the void, the maximum pressure calculated would still only be:

P (psia) =

(6.79E-6 moles + 7.92E-7 moles) (0.0821 ltr-atm/°K.mole) (6030 K)

(14.7 psi/atm) / 1.95E-5 liters P (psia) =

283 Pressurization experiments on this type of encapsulation conducted by ANSTO suggest that failure or can-rupture will not occur at pressures less than - 2000 psi (15 MPa) [4].

The potential pressurization mechanisms (gasses) are miniscule and even if totally released, would be safely contained in the encapsulation. This target and encapsulation type has been utilized at several reactors around the world and there have been no reported incidences of failure related to target pressurization. The irradiation container will satisfy the limitations of MURR Technical Specification 3.6.i.

(')Note:

As required by MURR Technical Specification 3.6.n, the maximum temperature limit on the LEU-foil target during irradiation is 3300 C (1/2 the melting temperature of aluminum).

10.

What are the electric power requirements during the LEU-foil target experimental run for the experiment instrumentation, cooling, reactor building ventilation, radiation monitoring, or other support equipment? Describe how the experiment is maintained in a safe condition upon a loss of normal power?

Electrical power to the experiment instrumentation, cooling, reactor building ventilation, radiation monitoring and other support equipment during normal reactor operation is supplied by the same source - the MURR Normal Electrical Power System.

Should a loss of normal electrical power occur, which will automatically cause a reactor shutdown, the Emergency Electrical Power System, through a 275-kW diesel generator, will provide electrical power to essential reactor equipment, including the facility ventilation exhaust and radiation monitoring systems.

Electrical power to the experiment instrumentation will also be supplied by the Emergency Electrical Power System, through the Uninterruptible Power Supply System.

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Cooling to the experiment is provided by the Pool Coolant System. The Pool Coolant System contains temperature, flow and pressure monitoring devices that will automatically cause a reactor shutdown in the event of a loss or degradation of coolant flow to the experiment. Should a loss of normal electrical power occur, the reactor will automatically shutdown and the heat production from the LEU-foil target will decrease to about 6% of its operating value in three seconds, or approximately 50 watts, and 16.7 watts after 21 minutes. This amount of heat can easily be dissipated into the 20,000 gallon reactor pool by the natural convection flow path that is automatically established in the reflector region on a loss of forced pool coolant flow.

Temperature of the LEU-foil target will be maintained well below any failure limits.

References

[1] Reactor Utilization Request 431, "Uranium, Low Enriched Uranium Foil," June 25, 2008.

[2] MURR Hazards Summary Report, Addendum 1, Section 3.17, page 86, February 1966.

[3] ANSTO Safety Assessment, Attachment 13, "Supporting Calculations for LEU Annular Target Pressure Capacity for HIFAR Target and Canning Specifications", Section 3.1.

[4] ANSTO Safety Assessment, Attachment 2, Appendix S, page 4, "Summary of Pressure Tests on Mo99 Annular Cans for HIFAR Trial Irradiations: Experimental Verification of Finite Element Analysis that predict Mo99 Annular Can Failure due to Internal Fission Gas Pressure."

[5] MURR Technical Data Report No. 0112, "Production of Fission Product Mo-99 Using the LEU-Modified Cintichem Process: Assessment of Fission Gas Release During LEU-Foil Target Irradiation & Disassembly."

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